Simulations of coupled heat transport, oxygen diffusion, and thermal expansion in UO2 nuclear fuel elements

Mihaila, Bogdan; Stan, Marius; Ramírez, José Ramón; Zubelewicz, Aleksander; Cristea, Paul Dan

doi:10.1016/j.jnucmat.2009.09.007

Cited by 38 publications

(22 citation statements)

References 27 publications

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“…The composition and the thermochemical properties of nuclear fuel materials greatly affect their behavior under nominal or off-nominal operating conditions. Several important properties, such as thermal conductivity, solid-state diffusion, creep rate, or volatile fission gas (such as I, Te) speciation and release are directly correlated to temperature and oxygen partial pressure (or the O/M ratio) in the fuel 1,2 . As those two parameters are likely to vary noticeably both in space and time during the fuel cycle, significant changes in the macroscopic material properties as well as in the speciation of the corrosive volatile gases, responsible for cladding ruptures are expected 3,4 .…”

Section: Introductionmentioning

confidence: 99%

Experimental evidence of oxygen thermo-migration in PWR UO2 fuels during power ramps using in-situ oxido-reduction indicators

Riglet-Martial

Sercombe

Lamontagne

et al. 2016

Journal of Nuclear Materials

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The present study describes the in-situ electrochemical modifications which affect irradiated PWR UO 2 fuels in the course of a power ramp, by means of in-situ oxido-reduction indicators such as chromium or neo-formed chemical phases. It is shown that sub-stoichiometric irradiated fuels under temperature gradient such as that occurring during high power transients are submitted to strong oxido-reduction perturbations, owing to radial migration of oxygen from the hot center to the cold periphery of the pellet. The oxygen redistribution, similar to that encountered in Sodium Fast Reactors fuels, induces a massive reduction/precipitation of the fission products Mo, Ru, Tc and Cr (if present) in the high temperature pellet section and the formation of highly oxidized neo-formed grey phases of U 4 O 9 type in its cold section, of lower temperature. To our knowledge, it is the first time that such a phenomenon, likely to greatly impact the corrosive fission gas release, is experimentally attested for PWR fuels. The parameters governing the oxidation states of UO 2 fuels under power ramps are finally debated from a cross-analysis of our results and other published information. The potential chemical benefits brought by oxido-reductive additives in UO 2 fuel such as chromium oxide, in connection with their oxygen buffering properties, are discussed.

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Section: Introductionmentioning

confidence: 99%

Experimental evidence of oxygen thermo-migration in PWR UO2 fuels during power ramps using in-situ oxido-reduction indicators

Riglet-Martial

Sercombe

Lamontagne

et al. 2016

Journal of Nuclear Materials

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“…To this end, the initial condition module UO2PXOxygenIc.C has been coupled with the thermal material model for hyperstoichiometric fuel ThermalUO2PX.C and oxygen diffusion kernel OxygenDiffusion.C. These modules are implemented to replicate the model for coupled oxygen transport by Mihaila et al [21,22].…”

Section: Thermodynamic State Computations In Bison Using Thermochimicamentioning

confidence: 99%

Coupling of thermochemistry solver THERMOCHIMICA with MOOSE/BISON

Simunovic¹,

Besmann²

2015

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“…Physics‐based fuel performance models rely on thermodynamics to provide key inputs, such as chemical potentials and material properties, for phase transformation and continuum transport simulations. This approach has already been used to describe thermal and oxygen transport in urania . However, to best represent the chemistry in a reactor over time, fission product (FP) generation must be accounted for.…”

Section: Introductionmentioning

confidence: 99%

“…This approach has already been used to describe thermal and oxygen transport in urania. 1,2 However, to best represent the chemistry in a reactor over time, fission product (FP) generation must be accounted for. Since Pr is one of the highest yield FP's, the Pr-O binary is an important subsystem of high burn up fuel.…”

Section: Introductionmentioning

confidence: 99%

Thermodynamic Assessment of the Pr–O System

McMurray

2015

J. Am. Ceram. Soc.

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The Calphad method was used to perform a thermodynamic assessment of the Pr–O system. Compound energy formalism representations were developed for the fluorite α‐PrO2–x and bixbyite σ‐Pr3O5 ± x solid solutions while the two‐sublattice liquid model was used to describe the binary melt. The series of phases between Pr2O3 and PrO2 were taken to be stoichiometric. Equilibrium oxygen pressure, phase equilibria, and enthalpy data were used to optimize the adjustable parameters of the models for a self‐consistent representation of the thermodynamic behavior of the Pr–O system from 298 K to melting.

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Simulations of coupled heat transport, oxygen diffusion, and thermal expansion in UO2 nuclear fuel elements

Cited by 38 publications

References 27 publications

Experimental evidence of oxygen thermo-migration in PWR UO2 fuels during power ramps using in-situ oxido-reduction indicators

Experimental evidence of oxygen thermo-migration in PWR UO2 fuels during power ramps using in-situ oxido-reduction indicators

Coupling of thermochemistry solver THERMOCHIMICA with MOOSE/BISON

Thermodynamic Assessment of the Pr–O System

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